The present invention relates to a method of producing a material for use in nuclear reactor environments and, in particular, to a method of producing a cladding, particularly a cladding tube, for nuclear reactor fuel. In particular, the present invention relates to a method of producing a cladding tube for nuclear fuel (that is, a tube for cladding the nuclear fuel) which is made of a zirconium alloy, and which is adapted for use in light-water reactors and heavy-water reactors of either the boiling water type or the pressurized water type.
Because of their excellent resistance to corrosion and very small neutron-absorbing sectional area, zirconium-based alloys have been used for making cladding tubes for holding fuel for nuclear plants, and for making fuel channel boxes.
These structures are irradiated with neutrons within reactors for extended periods of time, and also exposed to water or water vapor at high temperatures and high pressures, so that zirconium oxide films are formed on the surfaces thereof as corrosion proceeds. Spots of white oxide also form on the surfaces thereof. These spots of white oxide, called nodular corrosion, grow with the progress of corrosion reactions, and often peel off.
The outer surface of a cladding tube, in particular, is corroded by water or water vapor at high temperatures, so that the thickness of the tube is reduced by nodular corrosion, and its strength decreases.
In order to prevent abnormal nodular corrosion, studies have been conducted into methods of total .beta.-quenching. It is known that, among the zirconium-based alloys, Zircaloy 2 (an alloy composed of zirconium containing about 1.5% tin, 0.1% iron, 0.1% chromium, and 0.05% nickel, all by weight) and Zircaloy 4 (an alloy composed of zirconium containing about 1.5% tin, 0.2% iron, and 0.1% chromium, all by weight) exhibit a markedly increased corrosion resistance if entire bodies made of them are heated quickly after a hot extrusion step, e.g., if both inner and outer surfaces of a cladding tube, for instance, are heated rapidly to the temperature region of the (.alpha.+.beta.)-phase or the .beta.-phase, and if they are quickly quenched (hereinafter referred to as total .beta.-quenching). (German Laid-Open Pat. No. 3,324,686).
A problem, however, occurs on the inner surface of a cladding tube for reactor fuel, as described below. The cladding fuel tube becomes brittle at an increased rate because of the interaction of the nuclear fuel and the products of nuclear fission, and also because of irradiation with neutrons, and hence tends to crack easily. This tendency is promoted by local mechanical stresses generated by differences in thermal expansion between the nuclear fuel and the cladding tube.
Stress corrosion cracking may develop on the inner surface of the cladding tube because of the presence of nuclear fission products such as iodine, cadmium, and the like that form during the operation of the nuclear reactor, and also because of simultaneous local stresses.
The inner surface of a cladding tube is also subject to the problem of stress corrosion cracking caused by gases (such as iodine) emitted by the combustion of the nuclear fuel making up the central core, and also by the expansion of the central core as it is sintered.
It is known to provide a layer of pure metal between the central core and the cladding tube in order to prevent the occurrence of stress corrosion cracking. In particular, a composite cladding tube has been proposed, in which the inner side of the cladding tube is lined with pure zirconium. (British Pat. No. 1,525,717).
The layer of pure zirconium has a thickness of 5 to 30% that of the cladding tube. Compared with a zirconium alloy, pure zirconium remains ductile during use, reduces local stresses that act on the cladding tube, and helps prevent the occurrence of stress corrosion cracking.
If the whole cladding tube is subjected to total .beta.-quenching, however, the resistance to nodular corrosion increases on the outer surface of the cladding tube in contact with the reactor water, but the inner surface of the cladding tube tends to be susceptible to stress corrosion cracking. This is considered to be due to the fact that the dendritic structure formed by the total .beta.-quenching is hard and has a low ductility. In addition, .beta.-quenched materials tend to be more susceptible to stress corrosion cracking than ordinary annealed materials which are not subjected to .beta.-quenching, even after cold-rolling and annealing.
If the whole composite cladding tube were subjected to total .beta.-quenching in order to increase its resistance to stress corrosion cracking, the inner surface of the cladding tube being lined with pure zirconium, solute elements of the Zircaloy, such as tin, iron, chromium and oxygen, diffuse into the pure zirconium when heated to high temperatures, and the SCC resistance (resistance to stress corrosion cracking) decreases.
It is known to provide a nuclear fuel element having a zirconium alloy cladding tube with improved corrosion resistance. The cladding tube comprises a metallurgical gradient across the width of the tube wall wherein the tube has a more corrosion-resistant metallurgical condition at the outer circumference and a less corrosion resistant metallurgical condition at the inner circumference. The metallurgical gradient can be generated by heating an outer circumferential portion of the tube to the high .alpha. or mixed (.alpha.+.beta.) range while maintaining the inner surface at a lower temperature followed by cooling of the tube. (hereinafter referred to as partial .beta.-quenching) (U.S. Pat. No. 4,576,654).
In above U.S. patent, the technique of partial .beta.-quenching is shown. Namely, the exterior of the tube shell within the induction coil was allowed to increase in temperature to about 1650.degree. F. (900.degree. C.), in about 22 seconds. The exterior temperature of the tube shell was held at about 1650.degree. F. (900.degree. C.), for about 8 seconds. The power was then cut off and the tube shell with water flowing through its interior was allowed to cool to about 400.degree. F. (204.degree. C.), in about 21 seconds.
The heat-treated tube shell was then reduced in a Pilger mill in three successive passes to a final cladding tube. After each reduction pass, the tube was annealed at about 1150.degree. F. (621.degree. C.), for about 2 hours.
In above U.S. patent, both of an outer surface and an inner surface of the cladding tube are held at same temperature, at above the recrystallization temperature (about above 830.degree. C.).
As the outer surface of the cladding tube is quenched from a temperature above the recrystallization temperature, the processing structure does not remain at the outer surface. Therefore the corrosion resistance at the outer surface is lowered and the nodular corrosion at the outer surface can not be prevented.
Besides, the inner surface of the cladding tube is quenched also at above the recrystallization temperature. Since intergranular precipitations are caused as a whole at the inner surface, therefore the SCC resistance (resistance to stress corrosion cracking) decreases.